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A key uncertainty in the design and development of magnetic confinement fusion energy reactors is predicting edge plasma turbulence. An essential step in overcoming this uncertainty is the validation in accuracy of reduced turbulent transport models. Drift-reduced Braginskii two-fluid theory is one such set of reduced equations that has for decades simulated boundary plasmas in experiment, but significant questions exist regarding its predictive ability. To this end, using a novel physics-informed deep learning framework, we demonstrate the first ever direct quantitative comparisons of turbulent field fluctuations between electrostatic two-fluid theory and electromagnetic gyrokinetic modelling with good overall agreement found in magnetized helical plasmas at low normalized pressure. This framework is readily adaptable to experimental and astrophysical environments, and presents a new technique for the numerical validation and discovery of reduced global plasma turbulence models., 13 pages, 5 figures

Understanding and predicting divertor heat-load width ${\lambda}_q$ is a critically important problem for an easier and more robust operation of ITER with high fusion gain. Previous predictive simulation data for ${\lambda}_q$ using the extreme-scale edge gyrokinetic code XGC1 in the electrostatic limit under attached divertor plasma conditions in three major US tokamaks [C.S. Chang et al., Nucl. Fusion 57, 116023 (2017)] reproduced the Eich and Goldston attached-divertor formula results [formula #14 in T. Eich et al., Nucl. Fusion 53, 093031 (2013); R.J. Goldston, Nucl. Fusion 52, 013009 (2012)], and furthermore predicted over six times wider ${\lambda}_q$ than the maximal Eich and Goldston formula predictions on a full-power (Q = 10) scenario ITER plasma. After adding data from further predictive simulations on a highest current JET and highest-current Alcator C-Mod, a machine learning program is used to identify a new scaling formula for ${\lambda}_q$ as a simple modification to the Eich formula #14, which reproduces the Eich scaling formula for the present tokamaks and which embraces the wide ${\lambda}_q^X{GC}$ for the full-current Q = 10 ITER plasma. The new formula is then successfully tested on three more ITER plasmas: two corresponding to long burning scenarios with Q = 5 and one at low plasma current to be explored in the initial phases of ITER operation. The new physics that gives rise to the wider ${\lambda}q_^{XGC} is identified to be the weakly-collisional, trapped-electron-mode turbulence across the magnetic separatrix, which is known to be an efficient transporter of the electron heat and mass. Electromagnetic turbulence and high-collisionality effects on the new formula are the next study topics for XGC1., Comment: 23 pages, 16 figures, submitted to Physics of Plasmas

Owing to its high magnetic field, high power, and compact size, the SPARC experiment will operate with divertor conditions at or above those expected in reactor-class tokamaks. Power exhaust at this scale remains one of the key challenges for practical fusion energy. Based on empirical scalings, the peak unmitigated divertor parallel heat flux is projected to be greater than 10 GW m−2. This is nearly an order of magnitude higher than has been demonstrated to date. Furthermore, the divertor parallel Edge-Localized Mode (ELM) energy fluence projections (~11–34 MJ m−2) are comparable with those for ITER. However, the relatively short pulse length (~25 s pulse, with a ~10 s flat top) provides the opportunity to consider mitigation schemes unsuited to long-pulse devices including ITER and reactors. The baseline scenario for SPARC employs a ~1 Hz strike point sweep to spread the heat flux over a large divertor target surface area to keep tile surface temperatures within tolerable levels without the use of active divertor cooling systems. In addition, SPARC operation presents a unique opportunity to study divertor heat exhaust mitigation at reactor-level plasma densities and power fluxes. Not only will SPARC test the limits of current experimental scalings and serve for benchmarking theoretical models in reactor regimes, it is also being designed to enable the assessment of long-legged and X-point target advanced divertor magnetic configurations. Experimental results from SPARC will be crucial to reducing risk for a fusion pilot plant divertor design.

SPARC is designed to be a high-field, medium-size tokamak aimed at achieving net energy gain with ion cyclotron range-of-frequencies (ICRF) as its primary auxiliary heating mechanism. Empirical predictions with conservative physics indicate that SPARC baseline plasmas would reach$Q\approx 11$, which is well above its mission objective of$Q>2$. To build confidence that SPARC will be successful, physics-based integrated modelling has also been performed. The TRANSP code coupled with the theory-based trapped gyro-Landau fluid (TGLF) turbulence model and EPED predictions for pedestal stability find that$Q\approx 9$is attainable in standard H-mode operation and confirms$Q > 2$operation is feasible even with adverse assumptions. In this analysis, ion cyclotron waves are simulated with the full wave TORIC code and alpha heating is modelled with the Monte–Carlo fast ion NUBEAM module. Detailed analysis of expected turbulence regimes with linear and nonlinear CGYRO simulations is also presented, demonstrating that profile predictions with the TGLF reduced model are in reasonable agreement.

We present a fully Bayesian approach for the inference of radial profiles of impurity transport coefficients and compare its results to neoclassical, gyrofluid and gyrokinetic modeling. Using nested sampling, the Bayesian Impurity Transport InferencE (BITE) framework can handle complex parameter spaces with multiple possible solutions, offering great advantages in interpretative power and reliability with respect to previously demonstrated methods. BITE employs a forward model based on the pySTRAHL package, built on the success of the well-known STRAHL code [Dux, IPP Report, 2004], to simulate impurity transport in magnetically-confined plasmas. In this paper, we focus on calcium (Ca, Z=20) Laser Blow-Off injections into Alcator C-Mod plasmas. Multiple Ca atomic lines are diagnosed via high-resolution X-ray Imaging Crystal Spectroscopy and Vacuum Ultra-Violet measurements. We analyze a sawtoothing I-mode discharge for which neoclassical and turbulent (quasilinear and nonlinear) predictions are also obtained. We find good agreement in diffusion across the entire radial extent, while turbulent convection and density profile peaking are estimated to be larger in experiment than suggested by theory. Efforts and challenges associated with the inference of experimental pedestal impurity transport are discussed., 38 pages, 19 figures, submitted for publication in Nuclear Fusion

Emission lines which appear in the spectral ranges of ground state transitions from n = 2 levels in He- and H-like argon ions are discussed. X-ray transitions from elements commonly found in tokamaks (tungsten, molybdenum, iron and sulphur) which radiate in the wavelength range from 3700–4000 mÅ are identified by comparison with atomic structure calculations. Individual lines from tungsten charge states in the vicinity of Zn-like W44+ are documented, along with B-like Mo37+. The behaviour of line ratios as a function of electron temperature is examined, in support of the identifications.

Results from recent experiment and numerical simulation point towards a picture of the L-H transition in which edge shear flows interacting with edge turbulence create the conditions needed to produce a non-zero turbulent Reynolds stress at and just inside the LCFS during L-mode discharges. This stress acts to reinforce the shear flow at this location and the flow drive gets stronger as heating is increased. The L-H transition ensues when the rate of work done by this stress is strong enough to drive the shear flow to large values, which then grows at the expense of the turbulence intensity. The drop in turbulence intensity momentarily reduces the heat flux across the magnetic flux surface, which then allows the edge plasma pressure gradient to build. A sufficiently strong ion pressure gradient then locks in the H-mode state. These results are in general agreement with previously published reduced 0D and 1D predator prey models. An extended predator-prey model including separate ion and electron heat channels yields a non-monotonic power threshold dependence on plasma density provided that the fraction of heat deposited on the ions increases with plasma density. Possible mechanisms to explain other macroscopic transition threshold criteria are identified. A number of open questions and unexplained observations are identified, and must be addressed and resolved in order to build a physics-based model that can yield predictions of the macroscopic conditions needed for accessing H-mode.

Two aspects of edge turbulence and transport in Alcator C-Mod are explored. The quasi-coherent mode (QCM), an edge fluctuation present in Enhanced Dα H-mode plasmas, is examined with regard to its role in the enhanced particle transport found in these plasmas, its in/out asymmetry, its poloidal wave number and its radial width and location. It is shown to play a dominant role in the perpendicular particle transport. The QCM is not observed at the inboard midplane, indicating that its amplitude there is significantly smaller than on the outboard side. The peak amplitude of the QCM is found just inside the separatrix, with a radial width 5 mm, leading to a non-zero amplitude outside the separatrix and qualitatively consistent with its transport enhancement. Also examined are the characteristics of the intermittent convective transport, associated with the larger scale turbulent structures, also called blobs, and typically occurring in the scrape-off-layer (SOL). These turbulent structures are qualitatively similar in L- and H-mode. When their perpendicular extent, occurrence frequencies and magnitudes are compared, it is found that their size is somewhat smaller in ELMfree H-Mode, while their frequency is similar. A clear difference is seen in the magnitude of these turbulent fluctuations in the far SOL, with ELMfree H-mode showing a smaller perturbation there than L-mode. As the Greenwald density limit is approached (n/nGW ≥ 0.7), blobs are seen inside the separatrix consistent with the observation that the high cross-field transport region, normally found in the far SOL, penetrates the closed flux surfaces at high n/nGW.

For steady state high-confinement-mode (H-mode) operation, a relaxation mechanism is required to limit build-up of the edge gradient and impurity content. Alcator C-Mod [Hutchinson et al., Phys. Plasmas 1, 1511 (1994)] sees two such mechanisms—EDA (enhanced D-alpha H mode) and grassy ELMs (edge localized modes), but not large type I ELMs. In EDA the edge relaxation is provided by an edge localized quasicoherent (QC) electromagnetic mode that exists at moderate pedestal temperature T 3.5, and does not limit the buildup of the edge pressure gradient. The q boundary of the operational space of the mode depends on plasma shape, with the q95 limit moving down with increasing plasma triangularity. At high edge pressure gradients and temperatures the mode is replaced by broadband fluctuations ( f

The guiding-center kinetic neoclassical transport code, XGC0, [C.S. Chang et. al, Phys. Plasmas 11, 2649 (2004)] is used to compute the heat fluxes and the heat-load width in the outer divertor plates of Alcator C-Mod and DIII-D tokamaks. The dependence of the width of heat-load fluxes on neoclassical effects, neutral collisions and anomalous transport is investigated using the XGC0 code. The XGC0 code includes realistic X-point geometry, a neutral source model, the effects of collisions, and a diffusion model for anomalous transport. It is observed that width of the XGC0 neoclassical heat-load is approximately inversely proportional to the total plasma current $I_{\rm p}$. The scaling of the width of the divertor heat-load with plasma current is examined for an Alcator C-Mod discharge and four DIII-D discharges. The scaling of the divertor heat-load width with plasma current is found to be weaker in the Alcator C-Mod discharge compared to scaling found in the DIII-D discharges. The effect of neutral collisions on the $1/I_{\rm p}$ scaling of heat-load width is shown not to be significant. Although inclusion of poloidally uniform anomalous transport results in a deviation from the $1/I_{\rm p}$ scaling, the inclusion of the anomalous transport that is driven by ballooning-type instabilities results in recovering the neoclassical $1/I_{\rm p}$ scaling. The Bohm or Gyro-Bohm scalings of anomalous transport does not strongly affect the dependence of the heat-load width on plasma current. The inclusion of anomalous transport, in general, results in widening the width of neoclassical divertor heat-load and enhances the neoclassical heat-load fluxes on the divertor plates. Understanding heat transport in the tokamak scrape-off layer plasmas is important for strengthening the basis for predicting divertor conditions in ITER., Comment: Submitted to Physics of Plasmas

Under most operational conditions of Alcator C-Mod, the dominant type of H-mode is the steady state enhanced Dα (EDA) mode, characterized by good energy confinement, continuously degraded impurity confinement and absence of regular edge localized modes (ELMs). In this regime, a quasicoherent (QC) electromagnetic mode (kθ~5 cm-1, f~100 kHz) is observed, localized in the region of the density pedestal. Experimental evidence suggests that the mode is responsible for enhancement of particle transport. It is shown experimentally that the QC mode can exist in a well defined region in edge temperature-safety factor space, favouring high edge q values (q95≥3.5) and requiring moderate pedestal temperature Teped ≤ Tc (Tc~400 eV). As edge temperature and pressure gradient increase, the QC mode is replaced by broadband low frequency fluctuations (f

The XGC1 edge gyrokinetic code is used for a high fidelity prediction for the width of the heat-flux to divertor plates in attached plasma condition. The simulation results are validated against the empirical scaling $\lambda_q \propto B_P^{-\gamma}$ obtained from present tokamak devices, where $\lambda_q$ is the divertor heat-flux width mapped to the outboard midplane and $\gamma_q=1.19$ as defined by T. Eich et al. [Nucl. Fusion 53 (2013) 093031], and $B_P$ is the magnitude of the poloidal magnetic field at outboard midplane separatrix surface. This empirical scaling predicts $\lambda_q \leq 1mm$ when extrapolated to ITER, which would require operation with very high separatrix densities $(n_{sep}/n_{Greenwald} > 0.6)$ in the Q=10 scenario to achieve semi-detached plasma operation and high radiative fractions leading to acceptable divertor power fluxes. XGC1 predicts, however, that $\lambda_q$ for ITER is over 5 mm, suggesting that operation in the ITER Q=10 scenario with acceptable divertor power loads could be obtained over a wider range of plasma separatrix densities and radiative fractions. The physics reason behind this difference is, according to the XGC1 results, that while the ion magnetic drift contribution to the divertor heat-flux width is wider in the present tokamaks, the turbulent electron contribution is wider in ITER. A high current C-Mod discharge is found to be in a mixed regime: While the heat-flux width by the ion neoclassical magnetic drift is still wider than the turbulent electron heat-flux width, the heat-flux magnitude is dominated by the narrower electron heat-flux., Comment: Submitted to Nuclear Fusion

Because the harmonics of toroidal Alfven wave eigenmodes have a different response to mass density at different points along a magnetic field line, the frequencies of these harmonics can, in principal, be used to infer the distribution of mass density along the field line. Here we present several improvements to the methodology and test our method using magnetic and plasma wave data from the CRRES spacecraft. Our method includes the calculation of toroidal frequencies in a Tsyganenko magnetic field assuming a polynomial expansion for the logarithm of the mass density with respect to a coordinate related to distance along the field line. We use a Monte Carlo distribution of frequencies about the observed peaks in order to infer an uncertainty for the mass density. The method only works well if the frequencies of the fundamental and second harmonic are known. We compare the inferred mass density at the spacecraft location with the electron density determined from the plasma wave experiment onboard CRRES. The observed electron density is about a factor of 2 lower than the mean of the inferred mass density for an ensemble of frequency combinations based on the uncertainties of the measured spectral peaks but is close to or within the error bars of the inferred mass density. In one of the cases examined, the inferred mass density had a local maximum at the equator, while in the other case the inferred mass density increased monotonically away from the equator.

High resolution measurements on the Alcator C-Mod tokamak [I. H. Hutchinson et al., Phys. Plasmas 1, 1551 (1994)] of the transport barrier in the “Enhanced Dα” (EDA) regime, which has increased particle transport without large edge localized modes, show steep density and temperature gradients over a region of 2–5 mm, with peak pressure gradients up to 12 MPa/m. Evolution of the pedestal at the L-H transition is consistent with a large, rapid drop in thermal conductivity across the barrier. A quasi-coherent fluctuation in density, potential, and Bpol, with f0∼50–150 kHz and kθ∼4 cm−1, always appears in the barrier during EDA, and drives a large particle flux. Conditions to access the steady-state EDA regime in deuterium include δ>0.35, q95>3.5, and L-mode target density ne>1.2×1020 m−3. A reduced q95 limit is found for hydrogen discharges.

The time-dependent radial and poloidal velocity of edge turbulence is evaluated using a 2D time-delayed cross-correlation analysis of fast camera data from the gas puff imaging (GPI) diagnostic viewing the outer midplane region in Alcator C-Mod. The local poloidal velocity fluctuations are averaged over the poloidal viewing region of the GPI diagnostic and radially resolved over ±2 cm around the separatrix. The resulting poloidal velocity usually has a broadband frequency spectrum in the range ∼1–20 kHz, and a radial correlation width which decreases with increasing line-averaged density. In some cases with ICRH heating there was also a coherent poloidal velocity oscillation at 6–7 kHz which was highly correlated with a poloidal magnetic field oscillation at the same frequency. Some of these results are similar to the geodesic acoustic modes and/or zonal flows described in previous experiments and theory.

Direction reversals of intrinsic toroidal rotation have been observed in diverted Alcator C-Mod Ohmic L-mode plasmas following electron density ramps. For low density discharges, the core rotation is directed cocurrent, and reverses to countercurrent following an increase in the density above a certain threshold. Such reversals occur together with a decrease in density fluctuations with 2 cm(-1)≤k(θ)≤11 cm(-1) and frequencies above 70 kHz. There is a strong correlation between the reversal density and the density at which the Ohmic L-mode energy confinement changes from the linear to the saturated regime.

Electron internal transport barrier (ITB) formation is observed during fully non‐inductive LHCD discharges on Alcator C‐Mod. The formation is characterized by an abrupt increase of the central electron temperature after a few current penetration time has passed following LHCD turn‐on. The ITB phase is often terminated by m = 2 mode activity. Kinetic equilibrium reconstruction and TRANSP/LSC analysis show shear reversal in the q profile during the ITB phase. TRANSP/LSC predictive analysis suggests that fBs∼60% is possible if a similar temperature gradient is maintained at r/a∼0.5 with H‐factor of 1.

International audience; Intrinsic rotation has been observed in I-mode plasmas from the C-Mod tokamak, and is found to be similar to that in H mode, both in its edge origin and in the scaling with global pressure. Since both plasmas have similar edge ∇T, but completely different edge ∇n, it may be concluded that the drive of the intrinsic rotation is the edge ∇T rather than ∇P. Evidence suggests that the connection between gradients and rotation is the residual stress, and a scaling for the rotation from conversion of free energy to macroscopic flow is calculated.

Experimental observations of LHCD at high density (ne>1020m∼3) on the Alcator C‐Mod tokamak are presented in this paper. Bremsstrahlung emission from relativistic fast electrons in the core plasma drops sharply in single null discharges well below the density limit previously observed on limited tokamaks (ω/ωLH∼2). Modeling and experimental evidence suggest that the absence of LH driven fast electrons at high density may be due to collisional absorption in the scrape off layer. Experiments show that the expected current drive density dependence is recovered for inner wall limited discharges across the range of densities scanned (0.5×1020 m−3

The pressure at the top of the edge transport barrier (or ‘pedestal height’) strongly impacts fusion performance, while large edge localized modes (ELMs), driven by the free energy in the pedestal region, can constrain material lifetimes. Accurately predicting the pedestal height and ELM behavior in ITER is an essential element of prediction and optimization of fusion performance. Investigation of intermediate wavelength MHD modes (or ‘peeling–ballooning’ modes) has led to an improved understanding of important constraints on the pedestal height and the mechanism for ELMs. The combination of high-resolution pedestal diagnostics, including substantial recent improvements, and a suite of highly efficient stability codes, has made edge stability analysis routine on several major tokamaks, contributing both to understanding, and to experimental planning and performance optimization. Here we present extensive comparisons of observations to predicted edge stability boundaries on several tokamaks, both for the standard (Type I) ELM regime, and for small ELM and ELM-free regimes. We further discuss a new predictive model for the pedestal height and width (EPED1), developed by self-consistently combining a simple width model with peeling–ballooning stability calculations. This model is tested against experimental measurements, and used in initial predictions of the pedestal height for ITER.

Recent results from the lower hybrid current drive experiments on Alcator C‐Mod are presented. These include i) MSE measurements of broadened LHCD current profiles; ii) development of counter rotation comparable to the rate of injected wave momentum; iii) modification of pedestals and rotation in H‐mode; and iv) development of a new FEM‐based code that models LH wave propagation from the RF source to absorption in the plasma. An improved antenna concept that will be used in the upcoming C‐Mod campaigns is also briefly described.

LHCD on Alcator C‐Mod is being used in plasmas with parameters similar to those expected on ITER for the purpose of tailoring the plasma current profile. LHCD experiments have also produced intriguing results related to momentum transport and edge pedestal physics that affect the toroidal rotation profile and the temperature and density profiles. Quantitative comparisons between local measurements and theory/simulation have been performed, confirming the off‐axis localization of the current drive, as well as its magnitude and location dependence on the launched n‖ spectrum and electron temperature. Applying LHCD during the current ramp saves volt‐seconds and delays the peaking of the current profile. Counter current toroidal rotation during LHCD has been observed in both L and H‐mode plasmas. In H‐mode plasmas the edge pedestal collisionality is reduced while the overall pressure in the pedestal increases slightly.

Nonlinear transfer processes between large-scale edge flows and the ambient broadband fluctuations have been shown to play a significant role in the dynamics of edge turbulence, including spreading power from coherent modes and suppressing turbulence at the formation of edge transport barriers. In order to predict thresholds of confinement regimes, both the transition dynamics and the parametric dependence of the nonlinear energy transfer must be studied. Since the expected flow damping terms depend on ion collision rates and local safety factors, recent experiments aimed also to explore the nonlinear drive at various values of the plasma current, density and amount of auxiliary heating. Nonlinear interactions between zonal flows and turbulence in L-mode are estimated using bispectral as well as time-resolved methods based on gas-puff-imaging in Alcator C-Mod [1].

The MIT Plasma Science and Fusion Center and collaborators are proposing a high-performance Advanced Divertor and RF tokamak eXperiment (ADX)-a tokamak specifically designed to address critical gaps in the world fusion research programme on the pathway to next-step devices: fusion nuclear science facility (FNSF), fusion pilot plant (FPP) and/or demonstration power plant (DEMO). This high-field (>= 6.5 T, 1.5 MA), high power density facility (P/S similar to 1.5 MW m(-2)) will test innovative divertor ideas, including an 'X-point target divertor' concept, at the required performance parameters-reactor-level boundary plasma pressures, magnetic field strengths and parallel heat flux densities entering into the divertor region-while simultaneously producing high-performance core plasma conditions that are prototypical of a reactor: equilibrated and strongly coupled electrons and ions, regimes with low or no torque, and no fuelling from external heating and current drive systems. Equally important, the experimental platform will test innovative concepts for lower hybrid current drive and ion cyclotron range of frequency actuators with the unprecedented ability to deploy launch structures both on the low-magnetic-field side and the high-magneticfield side-the latter being a location where energetic plasma-material interactions can be controlled and favourable RF wave physics leads to efficient current drive, current profile control, heating and flow drive. This triple combination-advanced divertors, advanced RF actuators, reactor-prototypical core plasma conditions-will enable ADX to explore enhanced core confinement physics, such as made possible by reversed central shear, using only the types of external drive systems that are considered viable for a fusion power plant. Such an integrated demonstration of high-performance core-divertor operation with steady-state sustainment would pave the way towards an attractive pilot plant, as envisioned in the ARC concept (affordable, robust, compact) (Sorbom et al 2015 Fusion Eng. Des. submitted (arXiv: 1409.3540)) that makes use of high-temperature superconductor technology-a high-field (9.25 T) tokamak the size of the Joint European Torus that produces 270 MW of net electricity.

In magnetically confined plasmas, the study of edge dynamics has many important implications which include overall confinement and performance, and, in particular, calls for a quantitative understanding of the particle inventory and its sources. In fact, when studying the particle, momentum and energy balance of the edge plasma, one must include the contribution of neutral particles. However, direct measurements of the neutral population, i.e. neutral density and associated ionization rate, are particularly arduous in that region. Based on a simple measurement of the excitation of neutral deuterium (and hydrogen), we installed a series of diagnostics in the Alcator C-Mod tokamak, from which we can obtain those two quantities with high radial resolution. In addition we installed a series of modified Penning- type gauges to measure the neutral pressure in close proximity of the discharge itself.

We develop and test a model, EPED1.6, for the H-mode pedestal height and width based upon two fundamental and calculable constraints: (1) onset of non-local peeling–ballooning modes at low to intermediate mode number, (2) onset of nearly local kinetic ballooning modes at high mode number. Calculation of these two constraints allows a unique, predictive determination of both pedestal height and width. The present version of the model is first principles, in that no parameters are fit to observations, and includes important non-ideal effects. Extensive successful comparisons with existing experiments on multiple tokamaks, including experiments where predictions were made prior to the experiment, are presented, and predictions for ITER are discussed.

Direction reversals of intrinsic toroidal rotation have been observed in Alcator C-Mod ohmic L-mode plasmas following modest electron density or toroidal magnetic field ramps. The reversal process occurs in the plasma interior, inside of the q = 3/2 surface. For low density plasmas, the rotation is in the co-current direction, and can reverse to the counter-current direction following an increase in the electron density above a certain threshold. Reversals from the co- to counter-current direction are correlated with a sharp decrease in density fluctuations with k R ⩾ 2 cm−1 and with frequencies above 70 kHz. The density at which the rotation reverses increases linearly with plasma current, and decreases with increasing magnetic field. There is a strong correlation between the reversal density and the density at which the global ohmic L-mode energy confinement changes from the linear to the saturated regime.

We report extended studies of the I-mode regime [Whyte et al., Nucl. Fusion 50, 105005 (2010)] obtained in the Alcator C-Mod tokamak [Marmar et al., Fusion Sci. Technol. 51(3), 3261 (2007)]. This regime, usually accessed with unfavorable ion B × ∇B drift, features an edge thermal transport barrier without a strong particle transport barrier. Steady I-modes have now been obtained with favorable B × ∇B drift, by using specific plasma shapes, as well as with unfavorable drift over a wider range of shapes and plasma parameters. With favorable drift, power thresholds are close to the standard scaling for L–H transitions, while with unfavorable drift they are ∼ 1.5–3 times higher, increasing with Ip. Global energy confinement in both drift configurations is comparable to H-mode scalings, while density profiles and impurity confinement are close to those in L-mode. Transport analysis of the edge region shows a decrease in edge χeff, by typically a factor of 3, between L- and I-mode. The decrease correlates with ...

The goal of the Lower Hybrid Current Drive (LHCD) system on the Alcator C-Mod tokamak [Hutchinson et al., Phys. Plasmas 1, 1511 (1994)] is to investigate current profile control under plasma conditions relevant to future tokamak experiments. Experimental observations of a LHCD “density limit” for C-Mod are presented in this paper. Bremsstrahlung emission from relativistic fast electrons in the core plasma drops suddenly above line averaged densities of 1020 m−3 (ω/ωLH∼3–4), well below the density limit previously observed on other experiments (ω/ωLH∼2). Electric currents flowing through the scrape off layer (SOL) between the inner and outer divertors increase dramatically across the same density range that the core bremsstrahlung emission drops precipitously. These experimental x-ray data are compared to both conventional modeling, which gives poor agreement with experiment above the density limit and a model including collisional absorption in the SOL, which dramatically improves agreement with experiment above the observed density limit. These results show that strong absorption of LH waves in the SOL is possible on a high density tokamak and the SOL must be included in simulations of LHCD at high density.

Recent experiments have led to a fundamental shift in our view of edge transport physics; transport near the last-closed flux surface may be more appropriately described in terms of a critical gradient phenomenon rather than a diffusive and/or convective paradigm. Edge pressure gradients, normalized by the square of the poloidal magnetic field strength, appear invariant in plasmas with the same normalized collisionality, despite vastly different currents and magnetic fields—a behavior that connects with first-principles electromagnetic plasma turbulence simulations. Near-sonic scrape-off layer (SOL) flows impose a cocurrent rotation boundary condition on the confined plasma when B×∇B points toward the active x-point, suggesting a link to the concomitant reduction in input power needed to attain high-confinement modes. Indeed, low-confinement mode plasmas are found to attain higher edge pressure gradients in this configuration, independent of the direction of B, evidence that SOL flows may affect transport and “critical gradient” values in the edge plasma.

Edge transport barrier (ETB) studies on the Alcator C-Mod tokamak [Phys. Plasmas 1, 1511 (1994)] investigate pedestal scalings and the radial transport of plasma and neutrals. Pedestal profiles show trends with plasma operational parameters such as total current IP. A ballooning-like IP2 dependence is seen in the pressure gradient, despite calculated stability to ideal ballooning modes. A similar scaling is seen in the near scrape-off layer for both low-confinement (L-mode) and H-mode discharges, possibly due to electromagnetic fluid drift turbulence setting transport near the separatrix. Neutral density diagnosis allows an examination of D0 fueling in H-modes, yielding profiles of effective particle diffusivity in the ETB, which vary as IP is changed. Edge neutral transport is studied using a one-dimensional kinetic treatment. In both experiment and modeling, the C-Mod density pedestal exhibits a weakly increasing pedestal density and a nearly invariant density pedestal width as the D0 source rate increa...

A circuit is described for use with Geiger-counter spectrometers which employ a second Geiger counter to monitor the X-ray output. It allows neither counter to record while either is quenched, and thus ensures that the ratio of counts recorded is equal to the ratio of quanta received and is independent of the quench times, the mean X-ray output and the X-ray output waveform. The same principle should be applicable to other methods of monitoring. The circuit described also suppresses the background counts for the fraction of the cycle during which a self-rectified X-ray tube produces no characteristic radiation.

It is well known that if a pure tone is presented binaurally so that its intensity is the same in each ear, and there is no phase difference between the notes at the ears, the resulting sensation is that of a source of sound located in the median plane and situated is that of a source of sound located in the median plane and situated either within the skull or outside it. If a difference of phase between the notes is gradually introduced without altering their intensities, and the frequency of the tone is, say, 800 cyc. /sec., then the source appears to move laterally towards the side of the ear in which the phase is made to lead. According to some observers, the movement is on a horizontal circular arc centred between the ears, while others consider it to be along a line joining the ears. There is general agreement among workers on localization that considerable difficulty exists in the detection of such movements at frequencies above 800 cyc. /sec., and that their extent is more limited, but they disagree as to the frequency above which the lateral motion ceases to be noted. Stewart (1920) obtained movement at 1280 cyc. /sec., but not 1536 cyc. /sec.; Banister (1925) found movements at 1040 and 1345 cyc. /sec., but anomalous in character. On the other hand, von Hornbostel and Wertheimer (1920) state that at about 800 cyc. /sec., the lateral movement no longer extends 90° left and right, but decreases as the frequency increases, and should reach zero at 17,000 cyc. /sec. Halverson (1927) found that observable right and left effects appear to be present up to the upper auditory limit. Below 1400 cyc. /sec. judgments of direction are fairly consistent, between 1400 and 3000 cyc. /sec. there is no particular difficulty if allowance is made for the more rapid onset of fatigue at these frequencies. Above 3000 cyc. /sec. lateral effects are still observable, though median localization is extremely difficult to achieve. Hartley (1919) gave theoretical curves showing the phase difference produced at the ears by sources of various frequencies at different distances from the head. According to these, above 650 cyc. /sec. the maximum movement of the apparent source would be less than 90°, though at 1860 cyc. /sec. there should still be about 27° of movement on either side of the median plane. These curves have been checked experimentally by Firestone (1930), who found for three frequencies the phase difference and amplitude ratio of the sounds entering the ears of a man-shaped wax dummy from a source at varying azimuths around the head and at different distances from it. The phase differences found were in good agreement with the values calculated by Hartley. Various methods have been used to present the notes to the ears and to vary the phase difference between them. In the experiments of Halverson the sounds were produced by a tuning fork and led to the ears through tubes, in one of which was incorporated a sliding section to vary its length and thus the phase difference at the ears. An objection to this is the possibility of resonances in the tubes, different on both sides on account of their different lengths, so that intensity changes might occur as well as changes oh phase. Stewart, and Banister, used notes generated electrically and presented to the ears by telephones. Continuous phase changes could be made with their generators, but the purity of the resulting notes was doubtful.

A dimensionless parameter dependence study of intrinsic torque has been performed on a database of H- and I-mode plasmas from the Alcator C-Mod tokamak. The torque was determined by comparing intrinsic angular momentum density profiles just before and just after L–H and L–I transitions. The intrinsic torque has been found to scale as β N 1.5 ρ * − 1.0 ν * 0.1 , with the parameter ranges 0.3 ⩽ β N ⩽ 1.5, 0.004 ⩽ ρ * ⩽ 0.011 and 0.04 ⩽ ν * ⩽ 0.9. Comparison with results from JET and DIII-D suggests that the intrinsic torque should be normalized by some measure of the device size. Depending upon this normalization, the estimated total intrinsic torques for ITER, SPARC and ARC are ∼20, ∼4 and ∼8 Nm, respectively.